This invention relates to a radioactive substance-removing apparatus used with a system of circulating a molten metal such as molten sodium applied as a coolant of, for example, a Fast Breader Reactor for the purpose of eliminating a radioactive substance from the molten sodium.
Hitherto, a cold trap has been used in purifying molten sodium conducted as a coolant through a circulating system used with, for example, a sodium-cooled nuclear reactor. The prior art process of removing impurities from the molten sodium is to decrease the temperature of the molten sodium and solubility of the impurities to crystallize them out, followed by filtration. The process is based on the principle that a decline in the temperature of the molten sodium reduces the solubility of the impurities contained therein.
Though very effective for elimination of impurities such as sodium oxide or sodium hydride from a sodium coolant, the conventional separator only using a cold trap presents difficulties in removing all impurities thereof. Namely where impurities are contained in the sodium coolant at such a low concentration as prevents them from reaching saturated solubility even when cooled, or the impurities originally have high solubility, then the impurities are not effectively, crystallized out, but simply pass unnoticed through a mass charged in the cold trap.
In such case, a radioactive substance is possibly dissolved in the sodium coolant at a higher concentration than allowed. Therefore, the known separator solely based on the cold trap fails to eliminate radioactive substances such as nuclear fission products (F.P.) or radioactive corrosion product (C.P.). To date, no method has been proposed which can effectively resolve difficulties encounted in the elimination of the above-mentioned radioactive substance.